Core Damage Research Articles

MODELING LOFW IN A PWR USING MELCOR

The Probabilistic Safety Assessment (PSA) is part of a Nuclear Power Plant (NPP) licensing process. It considers the elaboration and updating of probabilistic models that estimate the risk associated to the operation, allowing the risk monitoring from the design to the plant decommissioning, for both operational as regulatory activities. The PSA identifies those components or plant systems whose unavailability contributes significantly to the Core Damage Frequency (CDF) and to the Large Early Release Frequency (LERF) of radioactive material. Based on the PSA Level 1 results for a reference plant under design, the Analysis, Evaluating and Risk Management Laboratory (LabRisco), located in the University of São Paulo (USP), Brazil, started the analytical investigation of severe accident phenomena using the US Nuclear Regulatory Commission (NRC) MELCOR2.2 code – focusing on the qualification of a group of specialists who will subsidize a PSA Level 2 for the same plant. This PSA Level 1 shows that the accident with large CDF contribution is the Loss of Feed Water Accident (LOFW). Therefore, the initial objective of the investigation was to model the progression of severe accidents during a LOFW for the reference Pressurized Water Reactor (PWR) and to analyze the response of the plant under these accident scenarios. During the course of the hypothetical LOFW in the reference plant, hydrogen was generated – by a reaction between the high temperature steam water and the fuel-cladding inside the reactor pressure vessel (RPV) but not representing a serious threat to the RPV integrity.

MELCOR steady state calculation of the generic PWR of 40MWth

After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

Formulation and performance evaluation of polymer-thickened supercritical CO2 fracturing fluid

Compared with conventional water-based fracturing fluid, supercritical CO2 fracturing fluid has the characteristics of inhibiting water sensitive damage and improving flowback efficiency. In view of the problems of low viscosity, poor sand-carrying performance and large filtration loss capacity of pure supercritical CO2, a supercritical CO2 fracturing fluid system had been prepared: 5% polydimethylsiloxane +5% kerosene +90% supercritical CO2. The rheological property, sand-carrying property, filtration loss property, core damage property and wetting inversion property of the system were systematically studied through performance evaluation experiments. At 325 K, the viscosity of the system can reach 4.67 mPa s, which was 54.26 times higher than that of pure supercritical CO2. The performance evaluation showed that the system had a good rheological property. The sand-carrying capacity was improved by 38.17%, and the filtration loss coefficient was decreased by 71.71% at most. The interaction results between CO2 and reservoir showed that the system could effectively increase the porosity and permeability of the core. For ultra-low permeability sandstone core, the porosity was increased by 7.71%, the permeability was increased by 115.20%. While for shale core, the porosity was increased by 6.99%, permeability was increased by 102.44%. At the same time, the system can change the core wetted by water phase into a neutral wetted one which was favorable for oil and gas flow. This system is expected to be further studied in oilfield applications.

A multi-area fatigue damage model of composite honeycomb sandwich panels under three-point bending load

The fatigue damage of composite sandwich panels is divided into two parts: the face sheet damage and the core damage. Both the two kinds of damage cause the performance degradation of panels, however only the core damage results in the ultimate failure. In this paper, a fatigue damage model of the face sheet and the core is proposed and three-point bending tests are carried out to obtain the deflection evolution curves. With the proposed model, the fatigue damage can be predicted accurately, which proves the validity of the fatigue damage model proposed in this paper.

Severe accident studies on the efficiency of mitigation devices in a SFR core with SIMMER code

Sodium-cooled Fast Reactors (SFR) are investigated as future Generation IV reactor concepts. In SFR, the core configuration is not the most reactive during the nominal reactor operation, its geometry change or coolant voiding can induce a reactivity insertion. Within these considerations, mitigation devices should be implemented into the core in order to limit the thermal energy released into the fuel during a severe accident and thereby the possibility to induce mechanical loadings of the reactor structure when vaporized materials expand. Complementary safety devices called Transfer Tubes (TT) are studied in a French concept as an effective mitigation measure of severe accidents to reach a final reactor safe state, even in case of a failure of all shutdown systems including the passive shutdown safety rods. More precisely the TT, with their upper part located in and above the core, cross the core support structure and are linked to the core catcher zone situated in the main vessel bottom. Their purpose is to extract fissile materials from the core zone and secondly to favor the transfer of the molten fuel from the core region to the core catcher in case of severe core damage. The fuel discharge out of the core zone is necessary in order to lower the core reactivity, the relocation into the core catcher is necessary to enable the stabilization, the sub-criticality and the fuel cooling. An extensive work on severe accident calculations and phenomena identification related to fuel discharge through TT was performed. These studies represent the outcome of the work performed in 2014–2019, performed with collaboration of Japanese JAEA and MFBR and Framatome in the framework of the SFR project carried out by CEA. Calculations of fuel discharge were carried out with the mechanistic calculation code SIMMER and the main results are presented in this paper. Firstly, the whole unprotected loss of flow (ULOF) accident sequence was calculated with the SIMMER code. From these calculations, the power and flow evolution were set as boundary conditions for a more detailed model with six fuel assemblies around TT. This model was used to focus on key phenomena and to check impact of design parameter as local bottom restriction inside TT on fuel discharge way.

Deep learning schemes for event identification and signal reconstruction in nuclear power plants with sensor faults

An initiating event (IE) is an event that may lead to core damage in a nuclear power plant (NPP), and being able to identify an IE is crucial in determining what actions to take. This process requires examining the data from the widespread NPP sensors. Any fault in these sensors can be a severe issue that may defer the mitigation action that would bring the NPP back into safe shutdown conditions. In this study, we focus on addressing two issues in this regard: (1) event identification with sensor faults, and (2) faulty sensor signal reconstruction under an IE with the help of two deep learning (DL)-based schemes. Through the superior capability of DL in feature extraction, discriminant information can be extracted to enable valid event identification even when the faulty sensors have been discarded. Owing to the nonlinear nature of the DL network model, the process information for sensor reading generation and the interrelations among the sensors in the event recordings can be used to reconstruct faulty sensor readings without the need for deterministic trend removal. Results from detailed experiments using data generated by a simulator of the Maanshan NPP in Taiwan are provided to illustrate the efficacy of the proposed approaches.

Honeycomb core failure mechanism of CFRP/Nomex sandwich panel under multi-angle impact of hail ice

The rudder of an airplane is fabricated as a sandwich panel and can potentially be damaged by hail ice. The dynamic responses of a CFRP/Nomex sandwich panel with barely any visible core damage under multi-angle impact of an ice sphere at various velocities are researched experimentally and by using numerical simulations at the mesoscale. The results show that the honeycomb core represents three types of failure modes: wrinkling, fracture of cell walls, and debonding of cell walls at the interface of two-glued-paper-walls (TGPW) due to resin failure. These failure modes exist at specific impact velocities and angles. This paper proposes a mesoscale numerical modeling method of a honeycomb structure to represent the debonding failure at the TGPW interface. In addition, the mechanical properties of these failures are revealed, as the wrinkling of the cell walls are caused by buckling and the fracture of the cell walls and debonding of the TGPW interface are caused by out-plane and in-plane shearing of cell walls. The results showed that the relationship between the denting and impact kinetic energy under multi-angle impact is linear with respect to the impact angle. The effects of impact angles on the contacting forces on a rigid wall and sandwich panel are presented as the amount of peak impact forces affected by the target material type. However, the intensity of the peak impact forces is rarely affected by the type of the target material. The peak impact force of nonvertical impact is greater than that of the normal impact when the normal components of the impact velocities are the same. This is because the impact angle affects the distribution of ice scatter fragments during the impact, and this generates different pressure distributions on the face sheet of the panel.

A new slick water system for hydraulic fracturing in tight reservoir to enhance imbibition oil recovery

In order to give full play to the role of imbibition of capillary force and enhance oil recovery of ultralow permeability sandstone reservoir after hydraulic fracturing, the mixed water fracture technology based on functional slick water is described and successfully applied to several wells in oilfield. The core of the technology is determination of influence factors of imbibition oil recovery, the development of new functional slick water system and optimization of volume fracturing parameters. The imbibition results show that it is significant effect of interfacial tension, wetting on imbibition oil recovery. The interfacial tension decreases by an order of magnitude, the imbibition oil recovery reduces by more than 10%. The imbibition oil recovery increases with the contact angle decreasing. The emulsifying ability has no obvious effect on imbibition oil recovery. The functional slick water system considering imbibition is developed based on the solution rheology and polymer chemistry. The system has introduced the active group and temperature resistant group into the polymer molecules. The molecular weight is controlled in 1.5 million. The viscosity is greater than 2mPa·s after shearing 2h under 170s-1 and 100℃. The interfacial tension could decrease to 10-2mN/m. The contact angle decreased from 58° to 22° and the core damage rate is less than 12%. The imbibition oil recovery could reach to 43%. The fracturing process includes slick water stage and linear gel stage. 10% 100 mesh ceramists and 8% temporary plugging agents are carried into the formation by functional slick water. 40-70 mesh ceramists are carried by linear gel. The liquid volume ratio is about 4:1 and the displacement is controlled at 10-12m3/min. The sand content and fracturing fluid volumes of single stage are 80m3 and 2500 m3 respectively. Compared with conventional fracturing, due to imbibition oil recovery, there is only 25% of the fracturing fluid flowback rate when the crude oil flew out. When the oil well is in normal production, about 50% of the fracturing fluid is not returned. It is useful to maintain the formation energy and slow down the production decline. The average cumulative production of vertical wells is greater than 2800t, and the effective period is more than 2 years. This technology overcoming the problem of high horizontal stress difference and lack of natural fracture has been successfully applied in Jidong Oilfield ultralow permeability reservoir. The successful application of this technology not only helps to promote the effective use of ultralow permeability reservoirs, but also helps to further clarify the role of imbibition recovery, energy storage and oil-water replacement mechanism.

Оценка динамических нагрузок на стены и перекрытие подреакторного помещения РБМК-1000 при паровом взрыве

One of the conditions for the safe operation of a nuclear power plant (NPP) unit is a comprehensive design and experimental justification of its failure-free operation in all operating modes and limitation of accident radiation consequences, including those in the case of severe beyond design basis accidents. According to the nuclear power industry development plans in Russia, new NPPs equipped with RBMK-1000 reactors are not supposed to be constructed in the future. Although the assigned service life of RBMK-1000 based power units that remain in operation is close to expiration, these power units account for most of the electricity generation in the total amount of nuclear power capacities in Russia (about 40%); therefore, the relevant industry organizations have decided to extend their operation. This article analyzes the severe accident evolvement scenario at an RBMK-based NPP during the stage of severe core damage, in the course of which fuel-containing masses collapse into the subreactor space filled with water. Once fuel-containing masses emerge in the sub-reactor room, they come in interaction with the reactor base concrete. There is a potential danger of the concrete floor slab melting and the corium collapsing into the bubbler pool water. The main strategy foreseen for keeping the molten core within the reactor space boundaries involves decay heat removal from the reactor and cooling of the support metal structures by supplying water. However, the filling of the subreactor space with liquid may give rise to conditions under which vapor explosion can occur. The maximum dynamic impact applied to the RBMK-1000 subreactor room walls in the event of possible interaction between the molten corium and water during a severe beyond design basis accident is estimated. It is shown that when the corium melt interacts with a large amount of water in the subreactor room, the kinetic energy of the resulting water vapor is sufficient to cause significant destruction of the power unit building. When the water level in the subreactor room falls below one meter, the destruction hazard becomes less probable. The mass of hydrogen released as a result of the interaction is also estimated.

Hypervelocity Impacts on Honeycomb Core Sandwich Panels Filled with Shear Thickening Fluid

The use of honeycomb core sandwich panels filled with a shear thickening fluid (STF) as a component of spacecraft micrometeoroid/orbital debris (MMOD) shielding was investigated using hypervelocity impact (HVI) testing. Incorporating a STF into shielding has the potential to reduce damage to the core and the likelihood of back-side facesheet perforation in the event of a HVI. The sandwich panels tested consisted of 1.27 cm thick hexagonal aluminum honeycomb core bonded between 0.064 cm thick aluminum facesheets. The STF displayed a marked rise in viscosity with increasing shear rate above a critical shear rate. It was based on low molecular weight polyethylene glycol (PEG) and hydrophilic fumed silica. Sandwich panel target specimens filled with the STF were subjected to HVIs by 1 mm diameter stainless steel spheres at nominal temperatures of -80°C or 21°C with nominal impact velocities of 4.8 km/s or 6.8 km/s. Additional specimens filled with PEG only were also impacted for comparison.Visual inspections and X-ray computerized tomography were used to assess impact damage. All of the panels experienced perforation of the impacted facesheet, facesheet bulging, localized delamination, and the formation of a cavity in the damaged core. STF-filled panels sustained significantly less damage than PEG-filled panels. None of the STF-filled panels were completely perforated during impact. In contrast, one of the PEG-filled panels impacted at the peak velocity was perforated. The remaining PEG-filled panel sustained substantially more honeycomb core damage and facesheet-core delamination compared to an analogous STF-filled panel. Sandwich panels filled with the STF provide superior HVI mitigation in comparison to panels filled with a Newtonian fluid (i.e., PEG). These experiments show that incorporation of STFs into MMOD shielding components has the potential to dramatically improve the HVI penetration resistance over a broad range of impact velocities and temperatures.

Component unavailability model for PSA Level I applications considering aging and startup degradation: An Application to an Emergency Diesel Generator System

The paper proposes an unavailability model, which considers standby component degradation due to repeated startups, including not only planned surveillance tests, but also additional startups due to real demands and, in cases when the component is detected failed in the planned test, extra tests related to corrective maintenance. The model is applied to evaluate the effect of additional tests on Emergency Diesel Generators from a Nuclear Power Plant’s Onsite Power System. These additional tests are not considered in traditional reliability models. The results show that excessive Diesel Generator testing during its working life may lead to a significant increase of component unavailability with reflections on the core damage frequency at the end of its overhaul maintenance period. This increment could justify anticipation of Diesel Generator overhaul maintenance in one or two working cycles.

Estimation of fission product source terms for the SGTR accident of a reference PWR plant using MELCOR and MAAP5

A steam generator tube rupture (SGTR) is a type of containment bypass accidents that if core damage is not prevented during this accident progression, has the potential to lead to a significant release of fission products (FPs) to the environment through the faulted steam generator (SG) or other possible paths in the secondary system. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during an SGTR accident of a reference pressurized water reactor (PWR) plant, and also investigates the impact of dedicated accident mitigation actions on the results of interest through the established modeling approach. To consider the integral responses of the reactor, SGs, and containment to a severe accident initiated by SGTR, detailed modeling was performed for a broad spectrum of plant safety/mitigation systems and relevant phenomena influencing the evolution of the accident. As the modeling and simulation tools for the SGTR accident analysis, MELCOR 2.2 and MAAP5 were employed to compare the results of interest. The results of the analysis demonstrate that the release of FPs into the environment could be estimated more realistically as a result of the best practice modeling for the SGTR accident and also it could be greatly decreased through dedicated mitigation actions. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

Dynamics rate of fiber chemical etching: new partial removal of cladding technique for humidity sensing application

Sensor-based optical fibers require thinner cladding for enhanced interaction between the evanescence wave and the environment. Fiber Bragg grated fiber sensors relying on an axial strain require chemical etching to weaken the fiber allowing more sensitivity in terms of strain. Ease of sensor multiplexing can be achieved through etching the mid-section of the fiber which is rarely reported with a custom-made holder. This paper proposes a technique for etching at the middle section of a fiber with a custom fiber holder. Chemical etching of a single-mode silica fiber was done with 48% hydrofluoric acid (HF). Etching proceeds until the fiber was completely etched. The optical spectrum was recorded at an interval of 3 min. Microscope visuals indicated possible etching with the mount design until ∼10 μm at 45 min with the fiber intact compared to a longer etching time. A longer etching duration than 45 min results in the optical spectrum displaying a null spectrum, depicting core damage. Etching of the fiber follows a linear trend and the reaction rate was determined to be 1.699 μm min−1 experimentally. The proposed fiber holder and receptacle were capable of etching a fiber with minimal defects while maintaining the ease of fiber handling.

Development of a multi-unit seismic conditional core damage probability model with uncertainty analysis

Various studies about multi-unit probabilistic safety assessment (PSA) have been conducted following the Fukushima accident. In this paper, an intuitive and simple model is developed to supplement the disadvantages of huge models that integrate all units in the same site. Through this framework, pre-assessment can be made for multi-unit events, the results of which can assist risk-informed decision making in a relatively short time. The model deducts multi-unit seismic conditional core damage probability (MUSCCDP) with a Bayesian belief network, and by including seismic inter-unit correlation factors, the model can also represent the extent of simultaneous and identical core damage in multiple units at the same site due to an earthquake. The outcome of the model confirms the impact that seismic intensity and seismic correlation have on MUSCCDP. Moreover, it provides an importance analysis of the components in the units and represents which components are the most vulnerable in multiple units. As the uncertainty analysis of the input data for the model is conducted by changing the distribution of the variables, the results raise the necessity for further uncertainty analyses in multi-unit seismic PSA before accepting the PSA results.

A Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors

Abstract The nuclear safety approach has to cover accident sequences involving core degradation in order to develop reliable mitigation strategies for both existing and future reactors. In particular, the long-term stabilization of the degraded core materials and their coolability has to be ensured after a severe accident. This paper focuses on severe accident phenomena in pressurized water reactors (PWR) compared to those potentially occurring in future GenIV-type sodium fast reactors (SFR). First, the two considered reactor concepts are introduced by focusing on safety aspects. The severe accident scenarios leading to core melting are presented and the initiating events are highlighted. This paper focuses on in-vessel severe accident phenomena, including the chronology of core damage, major changes in the core configuration and molten core progression. Regarding the mitigation means, the in-vessel retention phenomena and the core catcher characteristics are reviewed for these different nuclear generation concepts (II, III, and IV). A comparison between the PWR and SFR severe accident evolution is provided as well as the relation between governing physical parameters and the adopted mitigation provisions for each reactor concept. Finally, it is highlighted how the robustness of the safety demonstration is established by means of a combined probabilistic and deterministic approach.

Development of a hybrid method to assess grid-related LOOP scenarios for an NPP

Loss of Offsite Power (LOOP) is the most dominant initiating event in the Core Damage Frequency (CDF) of Nuclear Power Plants (NPPs). Since the frequency of occurrence of LOOP depends on the grid configuration, outage schedule, load flow, etc., even a relevant generic frequency may result in significant uncertainty.In this study, an analytical hybrid methodology, which combines the probabilistic and deterministic modules, is introduced to systematically estimate the frequency of Grid-related LOOP. The proposed probabilistic module develops the relevant event-trees and fault-trees in SAPHIRE7 to identify the LOOP sequences upon the response of the distance protection system to a fault on the transmission lines. Meanwhile, the identified scenarios are confirmed based on the simulation results from the deterministic module in DIgSILENT-PowerFactory15.1.15-out-of-48 scenarios (100-out-of-300 sequences) are identified as grid-related LOOP upon faults on the four transmission lines of the test grid, and result in 1.76E-06/yr for the grid-related LOOP frequency. The developed methodology enables us to trace a LOOP chain, from a primitive presumed fault through the progress of the cascade. The grid-related LOOP scenarios in a certain network can be ranked probabilistically, with the desired level of details in both deterministic and probabilistic modules.

Internal structure of the San Jacinto fault zone at the Ramona Reservation, north of Anza, California, from dense array seismic data

SUMMARYWe image the internal structure of the San Jacinto fault zone (SJFZ) near Anza, California, with seismic data recorded by two dense arrays (RA and RR) from ∼42 000 local and ∼180 teleseismic events occurring between 2012 and 2017. The RA linear array has short aperture (∼470 m long with 12 strong motion sensors) and recorded for the entire analysed time window, whereas the RR is a large three-component nodal array (97 geophones across a ∼2.4 km × 1.4 km area) that operated for about a month in September–October 2016. The SJFZ at the site contains three near-parallel surface traces F1, F2 and F3 from SW to NE that have accommodated several Mw > 6 earthquakes in the past 15 000 yr. Waveform changes in the fault normal direction indicate structural discontinuities that are consistent with the three fault surface traces. Relative slowness from local events and delay time analysis of teleseismic arrivals in the fault normal direction suggest a slower SW side than the NE with a core damage zone between F1 and F2. This core damage zone causes ∼0.05 s delay at stations RR26–31 in the teleseismic P arrivals compared with the SW-most station, and generates both P- and S-type fault zone trapped waves. Inversion of S trapped waves indicates the core damaged structure is ∼100 m wide, ∼4 km deep with a Q value of ∼20 and 40 per cent S-wave velocity reduction compared with bounding rocks. Fault zone head waves observed at stations SW of F3 indicate a local bimaterial interface that separates the locally faster NE block from the broad damage zone in the SW at shallow depth and merges with a deep interface that separates the regionally faster NE block from rocks to the SW with slower velocities at greater depth. The multiscale structural components observed at the site are related to the geological and earthquake rupture history at the site, and provide important information on the preferred NW propagation of earthquake ruptures on the San Jacinto fault.

Assessment of management measures on station blackout accident in Unit-1 Bushehr nuclear power plant

The radiological impact of nuclear power plant accidents is of public concern. Therefore, the accident management in order to minimize these effects is vital. Knowledge of plant behavior under circumstances of Station Black-Out (SBO) is essential for safer accident management. Thus, in this paper, the SBO accident as well as the effects of management measures on the behaviors of the Bushehr Nuclear Power Plant (BNPP) parameters under SBO has been analyzed with consideration of three scenarios using the RELAP5 code. The results from the SBO simulations showed that in case 1, the fuel cladding temperature has risen to 1200 °C in 11,250 s, which is more than the allowed temperature. Operator actions in case 2 and 3, maintain the fuel cladding temperature in the range of the permissible level. Moreover, these results determine accessible time for the operator’s actions to prevent core damage under SBO conditions in the selected scenarios.

Thermal response and edgewise compression failure of thermally degraded sandwich composite structures

Manufacturing and repair tasks using additive and subtractive laser processing (e.g., paint stripping and drilling) are becoming widely used in ships, aircrafts, and bridge structures. As with fire and exhaust impingement, such low level heat fluxes (<10 W) could negatively impact mechanical properties of composite structures. In this work, thermal excursions were simulated via continuous-wave laser exposure of the center region of a composite sandwich construction (i.e., E-glass fiber-reinforced vinyl ester matrix composite facesheets with a PVC foam core). Sandwich specimens were investigated with different slenderness ratios, facesheet thicknesses, foam core thicknesses and type of core materials. The 3-D temperature history response was measured with the spatially calibrated IR camera technology to map out the resulting damage state across the thickness of the specimens. The morphology of the foam core damage and interfacial debonding were postulated with the temperature profiles and evidenced by the X-ray CT images. Depending on the slenderness ratio and the volume of material heated above Tg, the thermally-induced damage significantly affects the edgewise compression strengths and failure modes.

Probability subtraction method for accurate quantification of seismic multi-unit probabilistic safety assessment

Single-unit probabilistic safety assessment (SUPSA) has complex Boolean logic equations for accident sequences. Multi-unit probabilistic safety assessment (MUPSA) model is developed by revising and combining SUPSA models in order to reflect plant state combinations (PSCs). These PSCs represent combinations of core damage and non-core damage states of nuclear power plants (NPPs). Since all these Boolean logic equations have complemented gates (not gates), it is not easy to generate exact Boolean solutions.Delete-term approximation method (DTAM) has been widely applied for generating approximate minimal cut sets (MCSs) from the complex Boolean logic equations with complemented gates. By applying DTAM, approximate conditional core damage probability (CCDP) has been calculated in SUPSA and MUPSA. It was found that CCDP calculated by DTAM was overestimated when complemented gates have non-rare events. Especially, the CCDP overestimation drastically increases if seismic SUPSA or MUPSA has complemented gates with many non-rare events.The objective of this study is to suggest a new quantification method named probability subtraction method (PSM) that replaces DTAM. The PSM calculates accurate CCDP even when SUPSA or MUPSA has complemented gates with many non-rare events. In this paper, the PSM is explained, and the accuracy of the PSM is validated by its applications to a few MUPSAs.